Chain reacting system



E. FERMI ET AL June 3, 1.958

CHAIN REACTING SYSTEM 10 Sheets-Sheet 1 Filed Feb. 16, 1945 zo ssaudzlzo) Zc/ifnzesses:

IN V EN TORS ":zrz'co Fermi Miles' C. efere MTE.

June 3, 1958 Filed Feb. 16, 1945 E. FERMI ET AL CHAIN REACTING SYSTEM Sheets-Sheet 2 Wwf@ June 3, 1958 E. FERMI ET AL 2,837,477

CHAIN REACTING SYSTEM 10 sheets-sheet :s

Filed Feb. l5, 1945 INVENTORS.' 5777"1'60 Fr'mz' BY /yzies C. eI/ere# 6 5 0 45g @a 4 7 Fw 6 o YW o l 9 5 l i, v 0% D A 0, m a W June 3, 1958 E. FERMI ETAL CHAIN REACTING SYSTEM 10 Sheets-Sheet 4 Filed Feb. 16, 1945 FIE , mi T o m5 j m f uw wEWl/e ne 15 fzl o w y lo June 3 1958 E. V*Prim/n Erm.` 2,837,477

CHAIN REACTING SYSTEM,

10 sheets-'sheet 57 Filed Feb. 16, 1945 VEN'ToRs: Fe m June 3, 1958 E. FERMI'ETAL 2,8372,477

CHAIN REACTING `SYSTEM Filed Feb. 1'6, 1945 y 10 Sheets-Sheet 6 Flc-3-1m.

Mgg, j,

Zl/Z rzesses:

June 3, 1958 E. FERMI ETAL CHAIN REACTING SYSTEM 10 Sheets-Sheet 8 Filed Feb. 1G, 1945 June 3, 1958 E. FERMI ET AL CHAIN REACTING SYSTEM 10 Sheets-Sheet 9 Filed Feb. 16, 1945 INYENTORSJ f7: rz'c o Ferl/zi M'Zes C efere `Pune 3, 1958 E. FERMI ET AL CHAIN REACTING SYSTEM 10 Sheets-Sheet 1,0

Filed Feb. 16, 1945 CHAIN REACTING SYSTEM Enrico Fermi, Santa Fe, N. Mex., and Miles C. Leverett, Oak Ridge, Tenn., assignors to the United States of America as represented by the United States Atomic Energy Commission Application February 16, 1945, Serial No. 578,278 Z Claims. (Cl. 204-193.2)

The present invention relates to the subject of nuclear fission and more particularly to a plant wherein the heat generated as a result of the fission process can be removed at a rapid rate and preferably in such a manner that it can be utilized for the production of power. In addition, the products resulting from the fission process in the plant can readily be removed without requiring complete dismantling of the plant. Y

The fission process is effected in a structure commonl referred to as a chain reacting pile orreactor which may comprise a plurality of bodies containing natural uranium geometrically arranged in an efficient neutron slowing material, such as carbon in the form of graphite.

Natural uranium contains both uranium isotopes U2?5 and U238 in the ratio -of 1 to 139. Considerwthe chain reaction as starting with the fast neutrons originating by nuclear fission in a uranium body. These neutrons are set free with very high energy of the order of magnitude of one million electron volts Aaverage andare therefore not in condition tovbe utilized eficientlyrto create new thermal neutron fissions in U235 when itis mixed with a considerable quantity of Um, particularly as inl the case of natural uranium. The energies of the fission-released neutrons are so high that most of thev latter would tend to be absorbed by the U233 nuclei -or lost from the system, and yet the energies are not generally high enough for production of fission by more than a small fraction of the neutrons so absorbed. For neutrons of thermal energies, however, the absorptioncross-section of U235, to produce fission, rises a great deal more than the simple capture cross-section of U2`38; so that under the stated circumstances the fast fission neutrons, after they are created, must be slowed down to thermal energies before they are most effective' to produce fresh fission by-re'- action with additional U235 atoms. If a system vcan be made in which neutrons are slowed down without much absorption until they reach thermal energies andl then mostly enter into uranium ratherthan into any other element, a self-sustaining nuclear chain reaction can be obtained, even with natural uranium.' Light elements, such as deuterium, beryllium, oxygen Vor carbon, the latter in the Vform of graphite, can be `used as slowing agents. A special advantage of the `use of the light elements mentioned for slowing downfast fission neunons is that fewer collisions are required for slowing than Sis the case with heavier elements, and furthermore,"the above-enumerated elements have very small neutron capture probabilities, even kfor thermal neutrons. yHydrogen would be most advantageous were it not for the'fact that there may be a relatively high probability of neutron captured by the hydrogen nucleus. .Both lberyllium and deuterium are expensive and not readily available in large quantities. Carbon in the form of graphite is a relatively inexpensive, practical, and readily available agent for slowing fast neutrons to rthermal energies. Recently, beryllium has been made yavailable in suiciently large quantities for test as to suitability Vfor use as a'neutron .ad safes Paten' 0 .i 2,837,477 Y Patented June 3,- 19,518-

ice

' the Um isotope of the uranium, certain types lof physical structurev should be utilized for the most efficient reproduction of neutrons, as precautions must be taken to reduce various neutron losses and thus to conserve neutrons for the chain reaction if a self-sustaining System is to be attained.

The ratio of the number ofrvfast neutrons produced by the fssions, to the original number vof fast neutrons creating the fissions, in a system of infinite sizeusing specific materials is called the reproduction Vor kmultiplication factor of the system and is denoted by the symbol K. This factor may be determined experimentallyfor any particular geometry by the process described inta copending application of E. Fermi, Serial No. 534,129,

Y filed May 4, 1944, now Patent No. 2,780,595. If K ca-n be made sufficiently greater thanunity .to createv'a net gain in neutrons Yand the system made sufficiently large so that this gain is not yentirelyulost by leakage from the exterior surface of the system, then a-self# Y sustaining chain reacting system can be built to produce Y exponentially. Such a rise will continue indefinitely if not controlled at a ldesired desired power out-put.

It is therefore the main object of our invention to construct a system comprising uranium and a slowing medium so `that neutron losses are reduced to such an extent that a controllable self-sustaini`ng neutron chain fission reaction is'obtained therein, with resultant regulated production of neutrons, liberation of powerfin the form of heat` and other forms, the production ofira'dioe active fission products and new elements both'radio active and stable producedrby the absorption lof neutrons.

During the interchange of neutrons in a systeni com prising bodies .of uranium of any size in a slowing mediun, neutrons may be lost in four ways; that iis, by absorption 1n the uranium metal or compound Without producing fission, by absorption in the slowing down material, by

Vdensity corresponding `to a Vabsorption in impurities present vin the system and by leakage from the system. These losses will be considered in the order mentioned. p l l Natural uranium, particularly by reason of its 'U2s content, has an especially strong absorbing power for neutrons when they yhave been slowedV down to moderate energies. The absorption in uranium at these energies is termed the uranium resonance absorption or capture. It A1s caused by the isotope (1238 and does notresult `in fission -but creates the isotope U239.which by two successive beta emissions forms the relatively stable nucleus 94239. Itis not to be confused with absorption or cap# ture of neutrons by impurities, referred to later. Neu# tron resonance Vabsorption in uranium may take place either on the surface of the uranium `bodies,"in which lcase the absorption is known as surface resonance abaU238 nucleus is Vany greater when the nucleus is ,at'rhe surface v'of a body of metallic or combined uranium, but

)f certain particular energies is inherently so high that zractically all neutrons that already happen to have those energies, called resonance energies as explained above, 1re absorbed almost immediately upon their arrival in he body of uranium metal 'or'uranium compound, Aand hus in effect are absorbed at the surface of such body. V'olume resonance absorption is due to the fact thatV some ieutrons make collisions inside the uranium bodyA and nay thus arrive at resonance energies therein. After suc- :essfully reaching thermal velocities, about 40 percent )fthe neutrons are also subject to capture by U238 with- )ut-fission, to produce U239 and eventually 94239.

It is possible,by proper physical arrangement of the materials, to reduce substantially uranium resonance abzorption. By the use of light elements as described tbove vfor slowing materials, a relatively large increment )f energy loss is achieved in each collision and thereore rfewer collisions are required to slow the neutrons o thermal energies, thus decreasing the probability of a ieutron being-at a resonance energy as it enters a uranium ltom. During the slowing process, however, neutrons Ire diffusing through the slowing medium over random aths and distances so that the uranium is not only ex nosed to thermal neutrons but also to neutrons of en- :rgies varying between theemission energy of fission and hermal energy. Neutrons Yat uranium resonance en- :rgies will, if they enter uranium at these energies, be tbsorbed on the surface of a uranium body whatever its iize, giving rise to surface absorption. Any substantial 'eduction of A.overall surface of the same amount of xranium relative to the amount of slowing material (i. e. he amount of slowing medium remaining unchanged) vill reduce surface absorption, and any such reduction n surfaceabsorption will release neutrons to enter di- .'ectly into the chain reaction; i. e., will increase the number of neutrons available for further slowing and thus for reaction with 11235 to produce ssion.

For a given ratio of slowing material to uranium, surace resonance absorption losses of neutrons in the uranium can be reduced by a large factor from the losses occurring ina mixture of line uranium particles and a slowing medium, if the uranium is aggregated into substantial masses in which the mean radius is at least 0.25 :entmeter for natural uranium metal and when the mean spatialrradius of the bodies is at least 0.75 centimeter Eor the oxide of natural uranium (U02). An important gain is thus made in the number o f neutrons made directly available for the chain reaction. is made when the uranium has more than the natural :ontent of iissionable material. Consequently, where it is desiredto secure a maximum K, we place the uranium inthe system -in the form of spaced uranium masses or bodies of substantial size, preferably either of metal, oxide, carbide, or combinations thereof. The uranium bodies can Vbe in the form of layers, rods or cylinders, :ubes or spheres, or approximate shapes, dispersed throughout the graphite, preferably in some geometric pattern. The term geometric .is used to mean any pat- `:ern or arrangement wherein the uranium bodies are dis- :ributed inthe graphite or other moderator with at least :ither a roughly uniform spacing or with a roughly sys :ematic non-uniform spacing, and are at least roughly miform in size and shape or are systematic in variations )f size or shape to produce a volume pattern conforming `ola roughly symmetrical system. If the pattern is a 'epeating or rather lexactly regular one, a system em- )odying it may be conveniently described as a lattice :tructure. Optimum conditions are obtained with natural lranium by using a lattice of metal spheres.

The number of neutrons made directly available to he chain reaction by aggregating Athe uranium into sepa- 'ate bodies spaced through the slowing medium is a :ritical factor in obtaining a self-sustaining chain reacion utilizing natural uranium and graphite. Th K fais used.

tor of a mixture of fine uranium particles in graphite, assuming both of themv to be theoretically pure, would only be about .785. Actual K factors as high as 1.07 have been obtained using aggregation of natural uranium in the best known geometry. With completely pure materials and uranium aggregates shaped as spheres it is possible to obtain K factors as high as 1.10 with a carbon moderator, 1.18 with a beryllium Amoderator and 1.3 with D20 as a moderator.

Still higher K factors can be obtained by the use of aggregation in the case of uranium having more than the naturally occurring content ofssionable materials such as U233, U235 or 94239. Adding such issionable material is termed enrichment of the uranium.`

It is thus clearly apparent that the aggregation of the uranium into masses separated in theslowing material is one of the most important, if not the most important factor entering into the successful construction of a selfsustaining chain reaction system utilizing relatively pure natural uranium in a slowing material in the bestrge-` ometry at present known, and is also important in obtaining high K factors when enrichmentV of the uranium The thermal neutrons. are alsorsubjcct to capture by the'slowing material. While carbon and berryllium have very small capture cross-sections for thermal neutrons, and deuterium still smaller, an appreciable fraction of thermal neutrons (about l0 percent of the neutrons present in the system under best conditions with graphite) A similar gain is lost by capture in the slowing material during diffusion therethrough. It is therefore desirable to have the neutrons reaching Ythermal energy promptly enter uranium.

In addition to the above-mentioned losses, which are inherently a part of the nuclear chain reaction process, impurities present in both the slowing material and the uranium add a very important neutron loss factor in the chain. The effectiveness of various elements as neutron absorbers varies tremendously. Certain elements such as boron, cadmium, samarium, gadolinium, and some others, if present even in a few parts per million, could prevent a self-sustaining chain reaction from taking place. It is highly important, therefore,'toremove as far as possible all impurities capturing neutrons to the detriment of the chain reaction from both the slowing material and the uranium. If Vthese impurities, solid, liquid, or gaseous, and -inelemental'or combined form, are present in too great quantity, in fthe uranium bodies or the slowing material or in, or by absorption from, the free spaces of the system, the self-sustaining chain reaction cannot be attained.v The amounts of impurities that may be permitted in aV system, Vvary with la number of factors, suchas the specific geometry of the system, and the form in which theuranium is used-that is, whether natural or enriched, whether as metal or oxide-and also factors such as fthe weight ratios between the uranium and the slowing down material, and the type of slowing down or moderating material used-for example, whether deuterium, graphite or beryllium. Although all of these considerations influence the actual permissible amount of each impurity material, it has fortunately been found that, in general, vthe effect of any given impurity or impurities can be correlated directly with the weight of the impurity present and with the K factor of the system, so that konwing the K factor for a given geometry and composition, the permissible amounts of particular imwww 'n i'atio of the system may be changed byfchangesin at:- mospheric pressure.' This eieet'mayV be eliminated'by enclosing or evacuating the systemV if desired. Ingeneral, the inclusion of combined nitrogen is 'to` be avoided.

The effect of impurities on the optimum reproduction factor K may he conveniently evaluated to a good ap proximation, simply by meansof certain; constantsknown as danger coeicients which are assigned to the various elements. These danger coeiicients for the impurities are each multiplied by the percent byweight of the. corresponding impurity, and the total sum of these products gives a value known as the total danger surn. This total danger sum is subtracted from the reproduction factor K as calculated for pure materials and for the specific geometry under consideration.

The danger coeflicients are deiined in terms of the ratio of the weight of impurity per unit mass of uranium and are based on the cross-section for absorption of thermal neutrons of the various elements. These values may be obtained from physics textbooks C111 the subject;y and the danger coeiiicient computed by theformula Ui Au n where ai represents the cross section for the impurity and au the cross-section for the uranium, Al the atomic weight of the impurity and Au the atmoic weight for uranium. If the impurities are in the carbon, they are computed as their percent of the weight of the uranium of the system.

Presently known values for danger coeci'ents for some elements are given in the following table, wherein the elements are assumed to have their mural isotopic4 constitution unless otherwise indicated, andA are convemently listed accordlng to thelr chemical symbols:Y

Element .Danger Element Danger Coetlieient Coetlcient 3 20V 1.8 4,6 0.61 2.7 1 is, 5 1.7' 2v 7.0 6 .3 2.5 2.5 16 6 82;

Where an element is necessarily used in an active part of a system, itis still to .be considered as an impurity; for example, in a structure where the uranium bodies consist of uranium oxides, the actual factor'K would ordinarily be computed by taking that fact intov account` using as a base K- a ValuecOmpUted for theoretically.

pure, uranium.. i

As a specific example, if thematerials ofthe system underconsideration have .0001 partbyfweght'off Co and Ag, the total' danger sum in K units for suchl an analysis would be:

.0001 17+.0001X 18:.0035 K unit This would be a rather unimportant reduction in the rel, y

production factor K unless the reproduction factor for a given system, without considering any impurities, is very nearly unity. If, on the other hand, the impurities in the uranium in the previous examplev had been.Li,Co,

and Rh, the total danger sum wouldY he:

.0310+.0017+.0050=.0377` K unit This latter reduction in the reproduction factor. for a given system would be seriousandmight well reduce` the reproduction factor below unity for certain geometries, so as to make it impossible to eiect a self-sustaining chain reaction with natural uranium and graphite, but

might still be permissible when using enriched uranium Yof neutrons. from the structure by leakage through the.

outer surfaces, which may overhalance the rate ofY neutron production inside the structure.l so tirata chain reaction. will not/be self-sustaining. v For eachjvalueof the reproduction' factor K greater than unity, there; thus afminimurnoverall size of a given structureknown as the critical.. Sizebove which the rate of.' loss` of neutronszby' dilusion tothe walls ofthe structureand` leakage, away from the structure is less thanthe r'ate of productionfof neutrons within the system. thus making the chainV reaction self- SJISai'ning-- The rate of diusion of, neutronsaway from a large structure in which they are being created through the exterior surface thereof may be. treated by mathematical analysis when thev value vof K andv certain other constants are known, as. the ratio of the. exterior surface. tothe volume becomesl less as the structureis enlarged.y

1n the case of a spherical structure employing uranium. bodies imbedded in graphite inthe geometries. disclosed herein andl withoutfan external reflector. thefollowingA formula gives the critical, over all. radiusA (R.) in feet: Y n

where a, b, and c are the lengths of the sides in feet. The critical size for a cylindrical structure is` given' by the formula, irrespective ofthe shaperof'the'uranium bodiesl I' cylinder height h ft. radius: R ft.

l-Iowever, when critical size is attained, by definition no rise in neutron density can be expected.l Itis'therefore necessaryto increasethesi/ze ofthe strcturefheyondj the critical' size; buty not to. theex-tent that the-period for doublingof! the neutron density is too short; as Vbe explained later. A desirable reproduction rati'ofor arr operatingstrUctureMith all control absonbersfremoved.

5.7 and l,at the ,temperature ofoperation is about 1.(105. The size at which this reproductionjratio can be obtained may be computedzfrom modicationsuof the .above formulae for critical size. For example, for spherical active struc tures the formula C K T:'R-z Y may be used to find 'R when K is known and r is the reproduction yratio and is somewhat over'unity. The same formula will, of course, give r for given structures for which K and R are known.

Critical size Vmay be attained with a somewhat smaller structure by utilizing a neutron reflecting medium surrounding the surface of the active structure. For example, a2 foot thickness of graphite having low impurity content, completely surrounding a spherical structure is eective in reducing the diameter of the uranium bearing portion by as much as 2 feet, resulting in a considerable saving of uranium or uranium compound.

The ratio of the average number of fast neutrons produced 'by fission of a `iissionable isotope to the average number ofthermal neutrons absorbed by the composition of whicbthe ssionable isotope is a component is a contant Afor any particular composition commonly referred to as the eta constant. Since fission of U235 produces about 2 f ast neutrons per fission the eta constant for pure U235 would be about 2 assuming that all neutrons produced were absorbed to produce ssion. However, U235 is usually ,used in conjunction with U238 and generally comprises but a very small portion of the composition subjected to treatment and accordingly this average constant is reduced because of absorptions by the composition by U238 which does not produce ssion. For natural uranium an average of about 1.32 new fast neutrons will be produced by iission of the nuclei of the U235 isotope and this value lis the eta constant for natural urabium. For iissionable materials other than natural uranium containing different concentrations of U235 and U238 or different iissionable isotopes such as 94239 the value o f the eta constant will be different.

The new fast `neutrons resulting from the iissions in the isotope U235 pass through the same cycle as just described, there being a certain portion which will pro- :luce fastiissiom some which will be lost to the chain reaction and others which will be slowed to thermal energy to be absorbed by uranium without iission.

By arranging the uranium in bodies or masses of suitable shape and suicientsize to minimize passage of neutrons having energies corresponding to or above resonance energies of H238 through the uranium preferably ina regular geometry resembling a crystal .lattice in chemical parlance and selecting the correct volume ratio of uranium to slowing material, and, further, by suitably limiting the impurities in the uranium and the slowing material and by building the structure to a proper size, it is possible to produce in each generation more fast neutrons by fission than are originally present to start the chain, so as to perpetuate the chain reaction in the ;ystem. As the chain reacting pile is then capable of producing more tast ssion neutrons at a greater rate than the rate at which neutrons are lost from the system, there would be an exponential rise in the neutron density, theoretically to infinity, as the pile is operated unless the rise'is controlled. Removable neutron absorbers, or impurities in the form of control rods, Vcan be used Vto control the exponentialrise beyond desired limits by inserting such rods lntothe pile. The point at which the exponential rise is stopped is a matter of choice, and, of course, will depend iponthe desired power output as well as considerations involving safety and eicieny of heat removal.v It is obvious that the rate of production of heat within the pile depend upon the operating neutron density in the system.` The;higher .the.density, the greater theA production r.18 of heat inthe system. :Moreover `the permissible power output of. thereactorisilimited Iby.the rate of heat removal and thereforethe permissible power output may be seriously reduced .where ;the.;coling media .is correlated through improper zones of the reactor. We have found that `about 92 percent of the total heat generated in a chain reacting system voriginates in the uranium, about 6 percent originates in the slowing medium, where graphite is used as such medium. VThe remaining 2 percent is generatedoutside the pile; that is, in the surrounding structures. Accordingly we' have found that maximum output of power may -be Vsecured by passage of the coolant into contact with Vor closely adjacent to the uranium bodies.

Following is a table showing more specifically the type and locale of the heat generated in the pile:

LSUMMARY BYIDYPE' lvl. e; v./ssion Percent Gamma radiation 23 11 Beta radiation 1l 6 Kinetic Energy of fission fragments 159 79 Kinetic Energy of neutrons 7 4 2. V.SUMMARY BYZLOCALE WHERE HEAT IS GENERATED M. e. vJssion Percent In Uranium 184 92 12 6 Outside Pile.; 4 2

3. SUMMARY BY TYPE AND LOCALE ln order to control thertemperature of the chain reaction and to prevent the accumulation of heat in the chain reaction pile, Ysome suitable circulating system must be employed to convey the heat away from the pile when a large power output'is desired.V The design of this system within the pile proper and the type of coolant employed are critical factors which, if not properly taken into consideration, will make it impossible to design a pile capable of producing a self-sustaining chain reaction.

The problem of removingrheat from a chain reacting system is complicated byvarious factors. The corrosive effect on uranium of most otherwise suitable circulating media is very troublesome. This factor is important primarily because of the presence in the system of high temperatures and intense neutron densities causing an acceleration of any normal' rate of corrosion. One of the most serious results Awhich may result lfrom the corrosiveaction of `a* circulating medium on the uranium is the physical deterioration of the uranium in the system. It is essential, then, that the circulating medium be of such-a char-V acter as not to destroy the uranium bodies in the system. Furthermore, many otherwise suitable cooling media absorb neutrons to such an extent that they cannot be used inthe pile.

. `VIn the presently described system, helium is used as the circulatinglnedium to remove heat from the structure. Helium .is an inertgasand therefore the corrosion problem with respect to the uranium is eliminated. Further- Imore, helium has practically no neutron absorptignfactor and is therefore very satisfactory from the po Wt* gf view of its effect on the reproduction of neutrons inthe chain reacting system. In other words thedan'ger cocicient of helium is very low. The danger Ycociliciexrt is delined in terms of the ratio of the weight o f impurityiin the chain reacting system per unit mass of uranium, and is based on the cross-section for absorption of thermal ncu-` trons. It therefore can be allowed to diffuse throughout the pile and can be circulated therein open channels without the use of tubes which might also absorb neutrons to the detriment of the chain reaction. s,

The fission products resulting from the chain reaction are, at least in general, highly radioactive. The diffusion ofthe fission fragments within the uranium is very slow, but fission fragments originating close to the surface of the uranium can escape into the helium 's gas, thereby `causing the gas to become radioactive. 'I'his, of course,

objectionable because it complicates the shielding problern in the circulating system outside the pile and requires steps to be taken to protect persons inthe vicinity of the gas ducts from the harmful effects produced byA this radio: activity. It may therefore be desirable to. cover or coat the uranium surfaces with metals such as aluminum or only a thin coating being required to prevent the flying fragments of the iissions from escaping from the solid uranium. This results in a substantial reduction in the radioactivity of the circulating helium gas, thus simi plif-ying the protection problem outside lof thepile.

In addition to the usual industrial hazards, personnel rrrust b e protected from injury by gamma rays Yand neutrons emanating from the reactor kand thecirculating gas and equipment, from injury by beta rays, or? by clse contact 'with radioactive materials,and from radioactive poisoning due to inhalation of radioactive gases emanating from the reactor. Y e

As one of the principal objects of the present invention, we provide a self-sustaining chain reacting power plant operating by virtue of nuclear fission at high'power'output in which temperature of the system is effectively controlled by efficient removal of heat therefrom'in an elective and safe manner at high power outputsofthe power plant. i

Other objects and advantages of ourv invention will become apparent from the following description and the drawings, which illustrate a power plant operating by virtue of 'nuclear fission -in which the heatI equivalent of 100,000 kilowatts is removed by circulation therethrough of 400,000, pounds of helium per hour.

Eig. l is a diagrammatic view of the reactor and the heat extraction system;

Fig. 2 is a vertical sectional view through the reactor showing in elevation the graphite and uranium lattice;

Eig. 3 is an enlarged, fragmentary, vertical sectional view through the center portion of the reactor shown in Fig. 2, indicating the relationship between the uranium `Figfl2 is a tsr planV View gian. assemblyzof, uranium A aletas whlcht im@ the Cartridges. shown in Fia v1,0 and '.lattice; FigiV l5 is a fragmentary, horizontal sectional Vview taken and graphite and further illustrating the structure at the bottom and top of the lattice;

Fig. 4 is a fragmentary, horizontal sectional view taken through the lattice on line 4 4 of Fig. 3 showing slightly moreA than one quadrant only, of the reactor;

Fig. 5 is a perspective view illustrating one arrangement for the graphite makingup the. lattice. structure;

Fig,V I6l isa plan View of a graphite brick shown in Fig. 5;

Fig. 7- is a horizontal sectional view taken on the line 7 1 of Fig. 6; Y

Fig. 8 is a transverse sectional view. taken onfthe line 8,--8 of Fig. 6;

Fig. 9 is a vertical sectionalvew through three bricks illustrating the manner in which bricks aresdoweledv togelber; 1

` Fig.' 10 is an enlarged sectionalvieyw through oneofv the gapliite uraniuml cartridges takenallogf tha( :gu/)T10 @Fia 11;, ,l

is. medal. the Center Qf the lattice Structure Where the heat is most intense; v

Eig- 13, is .an assembly Qi uranium Plates corresponding le Fig,- 12 but Whell is employed Semewhataway fram ...e tenter Q? the 'lattiss Strucwre Where. the heat generated islefss intense; Y u l i Fig. 14 is an enlarged, detailed, vertical sectional view through .a PQrfiOn 0f the reactor` galla@ `ltell Qf Fie-.2 showing therelationship between the lattiestrusture, the infellal Shield for the reatgranfl the guide tubes ,whla extend from. the graphite ,fllrsuslt the 'internal shield Providing a Passage' follaraias the. uranium iste. the

thfuah the lower Part (if the reader along line ,l5-15 of Fig.' 2 showing the arrangement of the valves use din discharging the' uranium from the lattice structure;

Fig. 16 is an enlarged, vertical sectional `view through one of the discharge valves on the line 16-16 of Fig, 15 showing the relationship of the valve to the lattice structure; 1

Fig. 17 is a horizontal sectional view takenon the line 17f17'of Fig. 16;

Fig. 18 is a vertical sectional view through one of the valves on line lsf-18, of Fig. lshowing thetrelationship of the valve to the lattice and the lattice support;

Fig. 19 is a rdiagrarrrrrratic view showingl thecontrol system for the power plant, theY electrical circuit being reduced to the lowest terms; f Fig. 20v is a longitudinal sectional View taken through gne ofthe filters' diagrammatically illustrated in Fig. l; Fig. 21 is a transverse sectional view taken on the line Fig'- 22 is an enlarged detail. ,Sectnal view Qf the fras-` vment marked 22 on the filter element shown in Fig. 20c

and

view showing a water cooled control rod.

Referring to Fig. l, theA neutron chain reaction iseff.

26 intoA heat exchangers 27, which may be of any well l known type to provide steam for the prime mover 27a. This steam may be used in any conventional manner-.to generate power as desired. The cooled helium then,

passes through pipe 28 into filters 29, which remove any solid matter from the helium, and thence into a battery of water cooled compressors 31 through a pipe 3.0. The compressors may be of the centrifugal or reciprocating type although the former is generally preferable. For4 most efficient heat removal, the heliumin the reactor `is maintained under pressure,V and for that reason the compressors 31` serve to establish and maintain they pressure and also serve as pumps to circulate the helium.

The high pressure gas leaves the compressors 3l through suitable'piping 32, and the heat resulting from ythe corn-r pression' is removed from the gas in after-coolers'331.'V

Frorn these coolers, the helium glas is returned to the reactor through piping 34. After-coolers 3,3 may bey used to` prcheat waterv to be turnedinto steam in exchangers 27. Similarly, the jacketsof the compressors 31y can be used to preheatV the feed water for exchangers 27. During operation the heat exchangers may kbecome more or less,

radioactive due to the radioactivity of the helium entering the exchanger. As a consequence cleaning of the exchanger may become didicult. Inforder toy minimige the necessity for cleaning it is found desirable, to use water treated` for reduction of its scale forming and corrosive` gfopertiesf inthe, heat exchanges t i 'm The pressure off-the helium gasvv enteringi the reactor i H Fig. 23 is an enlarged, fragmentary detail sectional 11 11S-pounds per square inch and the temperature is 120 degrees Fahrenheit. About 400,000 pounds ,ofggasgare circulated through thereactor "shown herein per hour. The gas leaves the reactor atfalpressure of 103.6 pounds per square inch and Vata temperature Tof -800 degrees Fahrenheit.

Referring to Fig. 2, the-reactor 25 consists of a lower chamber 35, which s ervesV as-a gas inlet and uranium discharge chamber. Abovethe chamber 4335 is a valve housing 36 which accommodates the uraniumdischarge valves presently to be described. The uraniumfgraphite lattice pile is contained in chamber V'37, above which is a hot gas-discharge chamber 38. A dome 39 completes the shell making up the reactor 25. These shell segmentsV are welded together so as to form one integral shell, and, as noted, the various segments are ofspherieal shape so as to give added strength to withstand the relatively Ahigh pressures existing in theshell. y

Referring to Fig. 3, the lattice structure comprises graphitecartridges 40 surrounded by graphite'bricks 41. The cartridges contain uranium in the form of cylindrical'aggregates or bodies 42 (see Figs. 10-13, inclusive). A high-grade AGOT graphite is preferably ernployed for the cartridges V*and the matrix lof the lattice structure. Surrounding the lattice is dead graphite 43 (Fig. 3), which need not be as puregnaphite asV that employed in the lattice. AGX grade graphite is satisfactory for use in the dead graphite area.

The graphite cartridges 40 are arranged in vertically disposed columns extendingfrom the bottom of the lattice to the top. The columns 'are disposed in parallel rows, as indicated in Fig. 4, so that the uranium in the graphite is arranged in a cubical lattice.

The graphite 41 may be in the form of bricks piled on top of each other, as shown iu^'-`ig.f5,V doweled together by means of dowel rods 41a passing through holes 411:. As illustrated, the bricks are22 inches long andV have a cross section `1l inches by.V 51/2`inches.Y Each brick finished but before drilling Weighs 65.8 pounds. Each brick is drilled to provide holes 44 through which the graphite cartridges '40 will pass. As the lattice structure is being built, the graphite b'ri'cks 41 are piled up so that the holes 44 Yare in alignment to form a continuous vertical passage 'fromfthe bottom to the top of the pile to accommodate each'of the columns of graphite cartridges. Sufficient space is provided between adjacent bricks and between the cartridges and the bricks to permit expansion of the graphite.

AEach of the graphite cartridges 40 (see Fig. 10) is provided with la longitudinal passage 45 extending throughout the entire height of the cartridge. Adjacent to the upper end of each cartridge is a cylindrical seat portion indicated at 47 provided with an annular shoulder 48 on which the uranium body 42 rests. `inasmuch as there is approximately only 50 percent free volume through the uranium bodies 42, the velocity of helium gas passing through the uranium is considerably greater than the velocity of the gas passing through the free passage in the graphite cartridges. i

' Most of the heat generated as a result of the neutron chain reaction is produced in the center portion of the lattice structure and progressively less heat is generated toward the outside of the structure. it is desirable that a greater amount of helium gas ,pass through the central portion of the lattice structure than through the outside portions. controlling the ow of the helium gas through the passages in the graphite cartridges is to provide a throat or constriction 46 (see Fig. 10) in the outer cartridges to create greater resistance to the ow of the gas through these restricted passages. In ,this manner, by providing the narrowest throat or constriction 46 in the outermost cartridges and further by gradually increasing the diameter of this constriction in the passages toward the center Aof ythe-lattice structure and lnally Vproviding no Because of this,

A satisfactory way forV -12 constriction inthe center portion vof the construction, it isjpossible to selectively=control the amount oliy gas ow through the va'u'ous'locationsA in the lattice to most electively remove heat from 'the system.

The average -ma-ss velocity of the helium passing throughlthe ducts in/ the lattice is about 5.56 pounds per second -persquare foot, while the maximum is 'at the centerot'thelattice Aand is approximately 12.8 pounds Vper-second per square foot.V The average linear velocity of the-gas in the ductl is about 110 feetper `second and thel maximum at the center of the lattice is about 254 feet per second. Thel vaverage mass velocity of the helium gas in theturaniumlelements 42 is 8.83 pounds per second per square toot, -while the maximum is about 20.4 pounds per VsecondA per-square foot. The average linear velocity of the Vgas in the elements is about 175 feet per second, Vwhileethe maximum is 404 Vfeet per second. On `an average, about 1068 pounds of helium gas per `hour pass through each column, the centermost columns ,conveying` the most, the .maximum for one column being approximately 2460 pounds per hour. The-heat transfer coetticient for the average column is about 1634'1B. t. -11.-s pervfhour pei-square foot per degree Fahrenheit, while 1 the coeflicient for the Y centermostr colum'nsis about 318 B. Vt.ru.s per hour per square foot per degreeFahrenheit. 400,000vpounds of helium are circulated per houryremoving -the heat equivalent of 100,000. kilowatts from the reactor. The power required for circulating .the helium :and running the auxiliaries of the plantv isvabout 12,000 kilowatts, leaving a net powerV ofxaround.88,000 kilowatts available to form steam :forexternal use.

, .The'uraniumbodies 42;are each made up of la plurality ofiparallelzplate's 49, disposedvertically and spaced apart with lugsV 50 serving as Aspacers-between the adjacent plates. In this manner, a plurality of vertical passages Sland 52` are Vprovided between the plates. The uranium disposed'invplate'rform asillustrated is'thus provided with a relativelylarge.amountof surface for cooling. The size ofea`ch ,uranium Vcylinder is about 3% inches by 3%' inches,`ahavingapproximately 50 percent free volume, and-each. Weighs about^13.4 pounds. This weight represents the value for uranium metal cylinders. If uranium carbide is used, the weight of each cylinder is about l4'.9'pounds.V .The overall mass ratio of graphite to uranium Vin the lattice is 5.4. The overall lattice structure vis inthe 4formof aicylinder 28 feet in diameter and 26 :feethigh andhas a two foot layer of dead graphite at the top 'and a layer ,on the'sides varying in thickness from l lfoot to 4feet. These gures `represent an operative lattice, but, of course, the invention is `not intended to be limited to this `specific example.

As shown in Fig. 3, a space isleft between the dead graphiteV layer .43 and the shell of the reactor and this spacegis lled with shredded asbestos indicated at 95 to a thickness ,of .'about one inch. `The! graphite will expand and contract as a result of the heat generated in the lattice and for this reason the space must be provided. l The asbestos layer will compress to permitthis expansionvbut. will serve to prevent leakage of Vhelium gas through this splaceso that the gas cannot by-pass the area of the reactor than elsewhere than in the region in the structure. Thisis'eiected, as illustrated in Fig. 12, passing the coolant in'- direct contact-with the uranium bodies and by decreasing the thickness of the uranium plates 49, fheiehv Providing e larger number of Plates with greater number of passages between them. inte rmedi` ate condition will existy in the lattice structure in positions between the center and the top or thevbottoiri. Thus, for these intermediate positions, uranium plates illustrated in Fig. 13 are employed wherein the thickness of the metal is greater than' that illustrated in Fig. l2 but not as great as that shown in Fig 11, so that the amount of cooling surface provided is somewhere between that shown in Figs. 1l and 12. Y

With the uranium 42 arranged in the form of vertically disposed plates 49, in the upper portion of the vertical Passage l-ii iS Seen. their e eohiiiiooosraseagef through the graphite and the uranium is provided throughout the entire length Vof each of the cartridges 4(1). As shcwn in Fig. 10,v each of the cartridges is charnfered, as indi. cated at S3, so as to facilitate the insertion of the cartridges into the openings 44 in the bricks.

The lattice structure lled to capacity usually contains about 12,060 uranium elements.

Referring again. to Fig. 2, the helium gas enters the reactor from the pipe 34 '(Fig. l) through intake nozzles 54, passes upwardly in the direction of the arrows through vertical passages in the lower layer 43a of dead graphite (Fig. 3) and'then continues upwardly through the passages 45 in the graphite eartridgeslt) (Figs. 3f andY throughout the entire' height of the" lattice, and na'lly through passages 56 in the upper layer 43 b 0f the dead graphite into discharge chamber .38. T he hotfgas leaves the reactor from the discharge chamber 38 throughV discharge nozzles 57. From the reactor, sas previously eirplained, the helium gas passes through the cooling circuit shown in Fig.'l.

At the top. of the graphite. 43 over each column of graphite cartridges 40 is a guide pipe 58(-Pigs. 2, 3, and 14.) socketed at itsv lower endfin the graphite at 5K9.

The top of the gas discharge chamber. 38 is bounded. by a Isteel floor 60 (Fig. 14) supported on a plurality of I-fbeams 61, which, in turn, are carried. b y the'steel` shell. Above the floor 60 is an internal neutron and gamma ray shield, generally indicated at 62, consisting of a 3 foot layer of graphite 63 and a layer of steel plates 64. The lowermfost steel plate 64a is one inch thick, and the next plate 64b is a one inch steel plate with about 2 percent to, l0;percent boron in the metal. The next 14 plates are of mild steel or. cast iron, and each'plate is about one inch thick.

The guide tubes 58 extend upwardly through the hot gas chamber 38 and through the graphite'layer 63; and steel layer 64 making up the internal shield 6,2, and terminate at the top of the steel layer 64. Slots'651are provided in the `guide tubes 5 8V throughout substantially the height of the hot gas chamber 3.8 through which, helium gas passing up through the lattice. is,v discharged. At the, top ofthe guide tubes 58, is a steel and graphite plug 66 consisting of a steel layer 66a corresponding in thickness to the thickness of. the steel layer 64 ofthe internal shield and a graphite layer 66h corresponding in thickness to the graphite layer 63v ofthe internal shield, A threaded rod 67 passes through the longitudinal center of the graphite portion 66.11 ofthe plug. and is threaded into the steel portiont 66a, as indicated at 68 A suitable nut 69 is threaded onto the, lower end of the rod 67 and awasher 70.` is disposed between the nut 6.9. and` the graphite 66ib engaging the lower facey of the graphite portion 66h of theplug. When `the rod is drawn up tight into the steel portion'66a of the plug,`the assembly becomes rigid. The. plug is provided with a llange 7.1 which rests on the upper surface of the internal radiation shield 62, and a screw eye '7.2 is threadedinto the top of the plug, providing means for, grasping the plug for withdrawal and insertion into the shield. i Y Y The dome 39 encloses a chamber above the steel and graphite plug 62. A covered manhole opening 73 is provided in the dome 39 for admitting persons in and out vvhieh 'is avertieal rassegne for Conveying the helium 65 top 81 -of the cars 59, the perforations SZ'being in*r .1.4 of, this Chamber for loading' titanium into. the lattice Ae' Previously ehpleiheel, theo'ieniuni iS mora., iii the. graphite oafirdges' 40, (Fiel0) ahdtheloadedeitid eQS ie lowered through the suidethhes 5S intothe' vert al passages S6 in theupper layer 43h of the dead graphite,` and finally down into the latticestructure. In order to prevent establishment of :a substantial concentration of radioactive material in the dome it is found desirable to establishl'a helium pressure within the chamber sonre-` vvhht higher than that. ih. eidioihins chambers iherehy'pre: venting or at least minimizing diffusion ofy nadioactiue gases info. the. dooie. i

Af, the, bottom of the' lattice, each of the ooloihns. ot graphite Cartridges is. siipporredfon. dead. graphite o'vlihder 74(1SeeFia- 3,) of AQX grade graphite thro ihio the. passages 'through the graahlfe. cartridges '441i The graphite. evlihliere; iii torn, 'are eopaorled onfdlhhp. valvesfgenjerallv 'ihdijeated iii Big5 2 at 52 There' is" plurality of such valves fer discharging the uranium from. thelam, Delio?, the operation of the present device, the trails.- Slllanic element 94 is produced, together with radif active fission products. After long. periods of operatiorh. the heoioh Rrodheie may So poisoh 'the materials ih` the deviee b3' neutron ohsorptioi' as to lower,l the replieguetieh ratio of 'the System! Ih orderto' perrietuafethe chain reaction, it is essential that thevalue` of the repro; ductionratio remain above. unity. Thus it may be de -v Sirahl'e to. remove the fission products from the laftiee.y from time to hine- Thie 'is done hv lremoving the uranium from the lhitieejehd, replacing if with, 4fresh material The radioactive iissionpro'ductsfand e1em`entf94 can then i be.. seperated from the uranium hv exhe'ofioh methode'- The redioeeiive heeioh Proditeie are useful ihmedieihe.

hlfe ih e hiehiher Similar to. Ugh?. may he heed' to eiirich natural uraniuml to increase its, eieiency in chain reacting systems, forA example. The .separation processes, formk nol partfof the present invention, sok that 4no purpose willbe served in describing herein the details thereof.

i is only resorted to after failure of the control andl safety rods, to act.v The dumping may be manual and/or autof matic upon rise of the neutron density in the, system to a dangerous level, as will be brought out later.

Each dump valve 5 9. Vin effect is a hat car mounted on 'wheel-s' operating onthe bases 77 of .adjoining I-beams 78, the bases serving as rails. The I-beams 78j arer'spaced' apart, as best shown in Eig. 1.8,'and on their; upper anges' support a floor 79, on which is, mounted` the 'dead graphite 43a (see Fig. 3). The .l-bearns 78,

.. in' turn, are supported on T-beams 5i?, whichare carried by the reactor shell.

4Referring to Figs. 16-18, inclusive, each dead graphite cylinder 74 supporting a column of uraniumpca'rtridge's 44) rests on the top 81 ol the dump valves or cars $97. directly over a plurality of perforations S2 through the alignment with the vertical passages through they dea d graphite cylinders 74. and the passages/SV through the column ofY graphite cartridges 413. Thus, with the dump valve ory car 59 disposed in the solid line position i shown in'Fig. 16, helium gas passes'upwardlyV fromthe inlet chamber 35 between the T-beams 8i), and Lbeamsj 78 through the perforations 82. in the top 31 of the valve` cars, and then through the vertical passages leadihaihrohsh the laftiee Shooters `The dump valves or cars 59 vary in length, depending upon their position in the lattice, as shown in Fig. 15,

the centermost car 59a being adapted to support 10 columns of cartridges, whereas the outermost car, shown at 59h, supports only two columns.

As illustrated in Fig. 15, the cross section of the reactor, insofar as the dmp valves are concerned, is divided into four quadrants, each quadrant being provided with V11 dump valves or cars and being served entirely independently of the other quadrants of the reactor. One quadrant only is completely shown in Fig. 15, though each of the other three quadrants is an exact duplicate of the one illustrated, except that in the left hand half of the reactor shown in Fig.V l5,V the dump valves or cars operate in the opposite direction from that shown for the right hand half of the reactor.

,Extending horizontally from the front of each of the dump valves or cars 59 is a dump rod 83 suitably fastened to the car as shown at 83a in Figs. 16 and 17. The dump rods 83 project through the shell of the reactor, as shown in Fig. 15, and extend horizontally to a valve control position (not shown) outside the reactor shell. Stuffing boxes, generally indicated at 84 inFig. l5, are provided around each of the dump rods 83 whereV the rod passes through the shell of the reactor to prevent escape of gases from inside the reactor shell. The dump rods 83 are slidable in guide tubes 85. Stuiing boxes 85a surround the guide tubes 85 where they pass through the concrete wall 96. The guide tubes-may be provided with suitable means (not shown) for absorbing or obstructing the path of neutron and gamma radiation therethrough as will be understood by those skilled in the art. Y

When dumping or discharging the uranium from the lattice structure for purposes of treating the uranium,all of the columns of uranium loaded graphite cartridges 40 supported on a single dump valve car are dumped or discharged simultaneously. In other words,freferring to Fig. l5, all 10 columns of uranium loaded graphite cartridges supported on the dump valve car 59a are discharged simultaneously. This is done by moving ythe car to the dotted line position 86 (see Fig. 16) wherein large circular openings 87, shown in Figs. 16 and 17, become disposed directly under the dead graphite cylinders 74, thereby permitting the dead graphite cylinders 74 and the column of uranium loaded graphite cartridges 40 supported thereon to discharge through the openings 87 in the dump valve car. The graphite and uranium thus discharged passes down into a downwardly extending chute 88 shown in Fig. 2, and accumulates at a gate valve 89. Upon opening of this gate valve 89, the uranium and graphite thus discharged can be passed into a suitable car or container.

A self-maintaining chain reaction, once started, must be controlled, for otherwise the neutron density may increase so rapidly that the reaction will reach violent proportions. The rate of heat generation in the lattice may exceed the rate of heat removal by the heat extraction system so that the temperature in the reactor will rise beyond a safet limit, even to the point of causing the uranium to melt with resulting break-down of the lattice structure.

Referring to Figs 2, 3, and 4, four control rods 90 and ve safety rods 91 pass vertically into the lattice through vertical openings 93 in the graphite 41'between the cylinders of uranium 42. These rods extend about two-thirds of the vertical distance through the lattice and serve to absorb neutrons so as to stabilize r stop the chain reaction.

The safety rods 91 are movable vertically-in the lattice but are to be disposed either in their lowermost position in the lattice or in their outermost position, `the latter position .being assumed when the chain 4reaction is in effect. The safety rods are lowered into the lattice only 16 l in the event of emergency or when the operation of the reactor is to be entirely stopped for any reason.

'Ihe control rods 90 also are movable vertically in the lattice but their positions will vary, their function being to stabilize the chain reaction and, with the help of the helium gas, to maintain a constant temperature in the lattice of about 800 F. as the maximum.

All of the rods and 91 may be controlled automatically, as will be explained later, or manually. Though nine rods in all are shown more or less rods may be used if desired.

The absorption of the neutrons by the control rods 90 is accompanied by considerable heat. Referring to Fig. 23, each control rod 90 may be in the form of a hollow tube closed at its lower end and surrounding an inner hollow tube 90a. The outer and inner tubesare spaced apart to form an annular passage 90b. Y Cooling water may thus be circulated through the control rod 90 entering the rod through a flexible tube 90C which conveys the water to the inside passage through the inner tube 90a, the water flowing downwardly through this tube to the bottom of the outer tube and then upwardly through the outer passage 90b finally leaving the rod through a second ilexible tube 90d. The Water may then be conveyed externally of the pile through a suitable cooling circuit (not shown) and then may be returned to the inlet tube 90C.

Extending above the lattice and surrounding each of the rods 90 and 91 is a guide pipe 94 (Fig. 3) to guide the movement of its corresponding rod. These pipes also may be provided with suitable means (not shown) for preventing escape of radiations from the reactor.

Referring to Figs. 2 and 15, a concrete shield 96 completely surrounds the reactor and extends vertically throughout the entire height of the reactor, terminating approximately eight feet above the topof the reactor dome 39. The concrete shield is cylindrical in cross section and is spaced a minimum distance of ten feet from the nearest point on the reactor shell. This concrete shield 96 is filled with water, shown at 97 in Fig. 2.

VExtending downwardly from the discharge nozzles 57 are discharge pipes 26a which are enclosed within the concrete .shield 96 and are disposed as closely as possible to the reactor shell so that there is a-substantial thickness of water between the discharge pipes 26a and the concrete wall 96. These discharge pipes extenddown to the bottom of the concrete shield to a position below the level of the ground indicated at 98, and are connectedA to themain discharge header 26, which likewise is disposed below the ground level.

Similarly, the gas intake header 34 (Fig. 2) is disposed below the level of the ground and passes through the concrete wall 96 of the shield enteringV the reactor through branch headers 34a, which are completely submerged in the Awater shield.

The construction of the lattice is commenced with the excavation for the foundations for the reactor shell proper and for the concrete structure or shield -containing the water lin which the reactor is immersed. The reactor shell itself, with its underground connections, dump valves, and graphite supporting beams is then built up to the level of the upper or internal radiation shield 62, and any necessary elevators and temporary super-structure required is erected. The graphite bricks 41, which have previously'been machined and bored, are then laid, doweling them together and taking particular care to keep the brick surfaces perfectly true and clean. When all the graphite bricks are laid, a temporary platform is laid over the top of the graphite bricks, the platform being provided with a hole directly over the location of each of the graphite cartridges. With a special reaming tool, each hole through the graphite is properly gauged throughout its length todetect the presence of any shoulders or projections, and such projections and shoulders are removed with the roaming tool. Following'this operation, the temporary platform is removed andthe beams 61 for supporting the upper or internalgraphite shield 62 are placed into position. The upper shield 62 is then built up of AGX or AGR graphite to its final thickness of about three feet. The beams for supporting the steel plates are then laid and the` 16 one-inch plates making up the steel portion 64 of the shield are laid on top of the beams. The guide pipes 53 are then inserted in place and suitable thermocouples (not shown) for measurement of temperature within the reactor adjacent the point at which the cooling uid is removed, installed on them. The steel and graphite plugs 66 are then inserted and the control and safety rod `guides installed. The piping for the helium gas is then installed and tted to the reactor shell and then the concrete shield 96 is poured. The mechanisms for operating the control and safety rods are finally installed in their proper place, and the entire unit is then in condition for operation except for the loading of the uranium.

From within the dome 39 of the reactor shell the graphite cartridges 40 loaded with the uranium plates 42v are lowered into the openings in the graphite. Several cartridges arranged in a column may be lowered into place at a time.

During the loading of the uranium, both the control rods and the safety rods 90 and 91 respectively are .disposed in their innermost position inthe lattice. As the charging of the uranium approaches the quantity necessary to produce an operating lattice, the loading operation will be suspended long enough for trial runs to be made. During the trial runs, the safety rods .are moved to their outermost positions and the control rodsare moved to various positions. Readings ofthe neutron density are made by the use of indium foils inserted in suitable testing slots in the lattice. The indium foil is inserted in the lattice and left there for a predetermined period of time, during which the foil is subjected to neutron bombardment. Then by suitable Geiger-Muller counters, the beta radiations from the indium foil radioactivity created by the neutron bombardment are counted. When it is found by this experimental means that the quantity of uranium in the lattice will support la chain reaction at the highest temperatures contemplated, the loading of the uranium is discontinued. All empty portions of the columnsare then filled with deadgraphite so Aas to complete the lattice structure. The" ratio lof graphite to carbon and the size of the uranium "plates may be determined in accordance with the principles discussed in copending application of Ferrnifand Szilard, Serial No. 568,904, filed December 19, 1944, Know-Patent No. 2,708,656.

The power plant above described is ideally adapted Y for automatic control to maintain the neutron density within the reactor substantially constant at a predetermined level, and thus give .a substantially constant power output. Due to the fact that large masses of materials are utilized in the reacting portion of the structure, there is a temperature lag therein. Consequently, it is convenient to monitor and control the reaction by means of ionization chambers which will measurethe neutron density at the periphery of the lattice Y portion ofthe structure. As the rate of neutron diffusion out `of a chain reacting system is always proportional to `the-rate of generation of neutrons within the structure,jtheioniza tion chambers can readily be placed atthe periphery of the pile or lattice, and in fact are preferably sopositioned in order that they be not subjected to the extremely high neutron densities existing within the lattice.

Before proceeding to a description of one type of control system that may be utilized, it is Vdesirable to point out the manner in which the control rods operate' to regulate the neutron density. In anyself-sustaining chain reacting structure adapted to produce power,.the.:neu-` 18 'tron multiplication ratio of the system must be greater "thanunity ',Forany Value over unity, by an amount suicient to overcome impurity losses in the system, the 'chain reaction becomes self-sustaining andthe neutron density Without control will increase exponentially in point of time, until the device is destroyed. For proper control, the system must be held in balance by maintaining'theA chain reaction at some point Where the production of new neutrons is balanced with the neutrons initiating the chain. Under these conditions, the reacting portion of the structure will 'continue to maintain the neutron density therein which obtained when. the system was balanced. l

However, in order to enable the reactor to reach adesired neutron density, the system must be permitted, for a vperiod of' time, kto rise in neutron density until the desired density is reached. After the desired density has been reached, it is Vnecessary .thereafter to hold the system in balance.

Inasmuch as the reproduction ratio of the lattice structure isj reduced by the presence of neutron absorbing impurities, such impurities can be introduced in the lattice in the form of the control rod which can be. of amaterialsuch as boron or cadmium, which will absorb large ess.

amounts of neutrons. The depth to which this control rod penetrates into the latticewill determine the reproduction ratio of the lattice and a range can be obtained between a conditiony providing a neutron reproduction ratio which is .greater than unity and acondition at which no chain reaction can be maintained. The exponential rise of neutron density can be made relatively fast or relatively'slow in accordance with whetherV the multiplication ratio is permitted to be much greaterthan one,'or only slightly greater `than one. There is a small percentage of delayed neutrons `emitted in the fission proc- 'Ifhese delayed neutronscausethe neutron density to rise in anappreciable time rather than almost instantaneously. The time for doubling the neutron .density increases as the multiplication ratio approaches unity, and byadjustment of the control. rods any desired rate of rise can be obtained up tothe maximum corresponding to the reproduction ratio characteristic of the given structure when allcontrol and safety rods arevremovved.

The vbroadmethod of control preferred is to withdraw allv safetyand control' rods fromthe structure; to `a point where :there 'is an exponential, and preferably slow, rise -in neutron density rwithin the structure. When a desired neutron density `has'been reached, the control rod isthen lreturned into the pile to a point Vwhere the reaction is balanced. This kbalance is then maintained to maintain a constant-power output in the reactor. The maintenance ofthe balance point with the control rod would be rela- `tively si-mplejwere it not for the fact that changes in temperature change the reproductionratio of the structure tern. --Such a method of control may be accomplished 'by automatic connection of the control Vrods with an ionization chamber measuring neutron density, positioned within the reactor close to the lattice.

p `Furthermore, due tothe exponential rise of .neutron density'within the reacting structure when Athe .multiplicationratio is greater than unity, all possible precautions must be taken to prevent a continued exponential rise in neutron density in case of failure of the -control rod to return to the balance position. It is for thisreason that .safety rods are provided. In case the neutronv density has risen'v to a very large gure'before the safety rods operate, f it mightthen be impossible for the safety rods to absorb a suicient number of neutrons to reducega dangerously `19 high neutron density to safe limits in a suflciently short period of time. Under these circumstances, there is no alternative but to dump the uranium bodies and thus destroy the lattice arrangement by which the self-sustaining chain reaction is made possible.

While there are many means by which the control rods, the safety rods, and the dumping can be operated, it is believed that by the illustration and description of one simplied circuit, other and fully equivalent circuits will be made apparent to those skilled in the art.

Referring therefore to Fig. 19, which shows diagrammatically and reduced to lowest terms one form of control circuit that may be used for regulating the output of the power plant hereinbefore described, and referring rst to control circuit A, a control ionization chamber 120 is placed within the reacting structure adjacent to the periphery of the lattice and filled with boron fluoride. A central electrode 121 is provided within the chamber 120 and connected to wire 122 leading outside of the reactor to a movable contact 123 on a resistor 125. Resistor 125 is connected across a relay coil 126. One side of relay coil 126 is connected to battery 127, the other end of which is connected to shield 129 around wire 122, Shield 129 is grounded, as is chamber 120. Alpha ray ionization due to neutron reaction with the boron within chamber 120 is proportional to the neutron density. Usual auxiliaries such as current ampliers, supplemented relays etc. may be employed if desired in order to increase the sensitivity of the device. Thus the current in resistor 125 is varied in accordance with neutron density reaching the ionization chamber. Relay coil 126 operates a relay armature 130, which is spring biased by spring 131 to contact one motor contact 133, and is urged by current in coil 126 to contact a second motor contact 135. Contacts 133 and 135 connect to the outside of split winding 140 of motor 141, the center connection 142 of which is connected through power mains 145 to armature 130. Motor 141 operates shaft 151), having on one side thereof a pulley 151, the other end thereof being connected through a magnetic clutch 152 to a control rod gear 154. Control rod gear 154 meshes with a rack S on a control rod 90. Pulley 151 has a cable 160 Wound thereon connected to a counter weight 161 so that the weight of the control rod is substantially balanced by counter weight 161, thus permitting motor 141 to run easily in either direction. As several control rods are to be utilized to control the pile, it is preferable that each rod be controlled by a separate ionization chamber and that the chambers be distributed around the periphery of the lattice. Y

Having described a circuit for controlling the position of a control rod, we shall now describe its operation, considering the safety rods withdrawn. Slider 123 on resistor 125, having previously been calibrated in terms of neutron density, is moved to the density position at which it is desired the reactor to operate, taking into account the difference in neutron density at the center of the lattice and at the periphery thereof during operation. This difference is a constant ratio at various operating densities. The reactor having at rest a neutron density much lower than the desired density at which relay coil 126 will receive enough current to operate armature 130, very little ionization takes place in ionization chamber 12), thus causing armature 130 to rest against contact 133. Motor 141 is thus energized to withdraw the control rod from the reactor to a point as determined by a limit stop 162, where the multiplication ratio of the reactor is iust suciently greater than unity to permit an exponential rise in neutron density with the reactor. The motor 141 will stall at stop 162, and should be of a type permitting stalling for the required time. TheV reaction at this position of the control rod becomes self-sustaining and the neutron density rises. In consequence, the ionization within chamber 12? rises. As the ionization in :hamber 120 increases, more and more current passes.

vrreached. Relay coil 126Y then operates to cause armatrol rods.

ture to connect with contact 135', thus reversing the motor 141 to drive the control rod into the reactor to a point where the neutron density starts to decay. The control rod 90 will thereafter hunt between a point above the balance position where the neutron density rises and a point below the balance position where the neutron l density decays, providing an average neutron density within the' reactor as determined by the setting of slider 123 on resistor 125.Y As the mass of the reactor causes any temperature change to lag behind any neutron density change, the temperature of the reactor is maintained substantially constant. `If desired, any of the well' known "anti-hunting circuits may be utilized, as will be apparent to those skilled in the art.V

The main purpose of the control circuit A is to regulate the control rods to substantially balance the neutron density to produce anyV desired average temperature within the reactor. i

Due to the fact that it might be possible for the control system as described to fail and thereby leave the control rods in a position where the neutron density would continus to rise indefinitely, both the safety rods 91 and the control rods 90 are preferably to be operated so as to enter the pile immediately upon any failure of the control rod system. One such emergency circuitv for operating the rods is illustrated in circuit B. The circuit for each of the safety rods is the same.

The-circuit B comprises a connection to main power line 145, one side of which leads through wire to the magnetic clutch 152 on the control rod drive. Wire 180, after passing through clutch 152, passes in series through a pressure switch 182m the helium input pipe 34 and then in a series with a thermostat 184 in the helium output pipe 26. Return wire 155 from the thermostat 184 then passes through a series of emergency switches 186, each of which being held closed by springs 187, and opened by manual operating handles 190. Thus, magnetic clutch 152 on the control rod motor shaft is in series with the power mains, the helium output temperature thermostat 184, the helium input pressure switch 182, and the emergency switches 136. The opening of any one of the switches will thus de-energize clutch 152 to remove the urge of the counter-weight fromthe con- The control rods'will fall by gravity into the reactor. Furthermore, any failure of power in supply line 145 will also de-energize clutch 152 and permit the control rod to fall into the pile.

Each safety rod 91 may be raised out of the pile by safety rod motor driving safety rod gear 191 through safety rod magnetic clutch 192, the safety rod being provided with a balancing counter weight 193 similar to that used on the control rod. Motor 190 is hand controlled by switch 194. Safety rod magnetic clutch 192 is connected in parallel with control rod magnetic clutch 152. Thus, upon any failure of power in the mains or the opening of any of the switches 182, 184, or 186, all of the rods will be relieved of their counter weights and will fall into the pile by gravity. Stops 195 may be provided on the safety rods, and any suitable braking action used to reduce the impact shock on the reactor.

However, in an abundance of caution, it is arranged that in case of failure of both the control and emergency circuits just described to stop the chain reaction, a portion of the lattice structure may be dumped by the use of circuit C. Dump valves 59 can be used for this purpose by connecting one or more of the dump valve rods 83 to a dump counter weight 216 urging the dump valve to open position. The dump valve is maintained closed by a dump valve latch 211 held in engagement with rack 212 on the dump valve rod83 by latch counter weight 214. Latch-211 may be withdrawn manually by means of handle 215, or electrically by means of coil 2,16, which is energized by a battery 217, which prefer- 

1. A DEVICE FOR EFFECTING A SELF-SUSTAINING NEUTRON CHAIN REACTION COMPRISING URANIUM BODIES GEOMETRICALLY ARRANGED IN GRAPHITE, A SHELL SURROUNDING THE GRAPHITE, THE SHEEL BEING SPACED SLIGHTLY FROM THE GRAPHITE TO PROVIDE AN EXPANSION SPACE FOR THE GRAPHITE, A QUANTITY OF HELIUM IN THE SHELL AND CIRCULATED UNDER SUPER ATMOSPHERIC PRESSURE THROUGH THE GRAPHITE AND IN HEAT EXCHANGE RELATIONSHIP WITH THE URANIUM, AND A RESILIENT PACKING IN THE SPACE BETWEEN THE GRAPHITE AND THE SHELL TO PREVENT PASSAGE OF HELIUM THERETHROUGH. 